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 Entered dateSiteRegionReactor typeEvent description
ENS 5436029 October 2019 13:58:00BraidwoodNRC Region 3At approximately 0945 CDT, on 10/29/2019, the Braidwood Station main control room was notified of the inadvertent actuation of 17 full sounding sirens affecting Braidwood Station in Will County, Illinois. Will County notified Exelon of the inadvertent actuation that occurred on 10/29/2019, at 0919 CDT, during the performance of regular maintenance on the siren equipment at the Laraway Communication Center at the Will County Sheriff Dispatch Center. This event is reportable per 10 CFR 50.72(b)(2)(xi), News Release or Notification of Other Government Agencies. The Braidwood NRC Resident Inspector has been notified. The siren equipment has been repaired and restored to service.
ENS 5428923 September 2019 14:15:00BraidwoodNRC Region 3

At 1106 CDT Braidwood Unit 1 experienced an automatic reactor trip due to lowering steam generator water levels following closure of the 1B steam generator feed water regulating valve.

The cause of the 1B steam generator feedwater regulating valve failing closed is unknown at this time and is under investigation.

Both trains of auxiliary feedwater started automatically following the reactor trip to maintain steam generator water levels.

All systems responded as expected with the exception of intermediate range nuclear instrument N-36 which was identified as being undercompensated following the reactor trip. Both source range nuclear instruments were manually energized in accordance with station procedures. Steam generator power operated relief valves lifted momentarily and reseated as designed in response to the secondary transient due to the reactor trip. The main steam dump valves are in service to the main condenser to provide heat sink cooling. The plant is being maintained at normal operating pressure and temperature. AC power is being provided by offsite power with the diesel generators in stand by and all safety systems available. There is no impact to Unit 2. This report is being made per 10 CFR 50.72(b)(2)(iv)(B) for a RPS actuation, 4 hour notification, and per 10 CFR 50.72(b)(3)(iv)(A) for an automatic actuation of the auxiliary feedwater system, 8 hour notification. The NRC Resident Inspector has been informed."

ENS 5409026 May 2019 01:58:00BraidwoodNRC Region 3At 1930 (CDT) on 5/25/2019, communications were lost with the main control room area radiation monitors. These detectors are used to determine if an emergency action level (EAL) has been reached for initiating condition RA3 (Radiation levels that impede access to equipment necessary for normal plant operations, cooldown, or shutdown). This unplanned loss of the ability to evaluate an EAL for initiating condition RA3 is considered a loss of emergency classification capability and is reportable as a Major Loss of Emergency Preparedness Capabilities per 10 CFR 50.72(b)(3)(xiii). This is an 8-hour reportable notification. Portable area radiation monitors have been established as a compensatory measure per station procedures. The NRC Resident Inspector has been notified."
ENS 537848 December 2018 06:12:00BraidwoodNRC Region 3

EN Revision Text: INOPERABLE CONTROL ROOM ENVELOPE Braidwood Station was performing Control Room Envelope Testing. During testing the Station identified a failed acceptance criteria. The Control Room Envelope is a single train system and could constitute a Loss of Safety Function. If a single train system is inoperable per Technical Specifications (TS), it is Reportable as a Loss of Safety Function per 10 CFR 50.72(b)(3)(v) regardless of the system's continued ability to meet the accident analysis requirements.

Both Units remain Mode 1, 100% power. The licensee will be notifying the NRC Resident Inspector. The acceptance criteria that failed was to maintain the control room pressure above the miscellaneous electrical equipment room pressure. The station has realigned ventilation to normal, and has entered TS Limited Condition for Operation (LCO) 3.7.10 condition B, which requires the station to restore to operable the control room envelope within 90 days or shutdown the plant. The station has also initiated contingency actions to verify SCBA (self contained breathing apparatus) are available and control room personnel are qualified to use SCBA.

  • * * RETRACTION ON 12/20/18 AT 1714 EST FROM ANTHONY SIEBERT TO JEFFREY WHITED * * *

On Wednesday, December 19, 2018, Braidwood Station concluded that the ENS notification 53784 could be retracted. It has been determined that the issue was not with the Control Room Envelope structure. Troubleshooting identified that the Unit 1 Upper Cable Spreading Room Area Supply Flow Control (OVC035Y) damper which supplies the Train A control room ventilation equipment room with air flow was not opening enough to supply the required flow. The subject duct work is shared by both A-train and B-train, and the flow through OVC035Y is controlled by a two-position actuator. The damper is less open when A-train is in operation (actuator energized) and more open when B-train is in operation (actuator de-energized). The only adjustments performed were to the actuator energized stroke limits which only affect the A train and thus a single train failure which could affect the safety function of both trains did not exist. Further calculations of unfiltered air inleakage into the Control Room Envelope (CRE) under a slightly negative differential pressure condition resulted in a calculated in leakage to the CRE of less than the maximum allowable unfiltered air inleakage for a radiological event of 436 scfm. The unfiltered air inleakage into the CRE assumed in the licensing basis analyses of Design Basis Accident consequences was never exceeded. Thus, TS Surveillance Requirement 3.7.10.4 continued to be met and entry into TS 3.7.10 Condition B was not required. The licensee has notified the NRC Resident Inspector. Notified the R3DO (Stone).

ENS 536528 October 2018 04:13:00BraidwoodNRC Region 3

On Monday, October 8, 2018 at 0111 CDT, during the initial containment entry for unit 2 refueling outage (A2R20), reactor coolant system pressure boundary leakage was discovered at the 2D Steam Generator bowl drain line. Unit cooldown to mode 5 is in progress.

This event is reportable under 10CFR50.72(b)(3)(ii)(A) for 'Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.'

The licensee has notified the NRC resident inspector. Approximately 0.1 gpm was leaking from the drain line. LCO 3.4.13 was entered and the licensee anticipates being in mode 5 within a couple of hours. The leak will be repaired prior to exiting the refueling outage.

ENS 534434 June 2018 14:00:00BraidwoodNRC Region 3At 0920 CDT, Braidwood Unit 1 reactor was manually tripped due to lowering steam generator water levels following a trip of the 1C main feedwater pump. The cause of the 1C main feedwater pump trip is unknown at this time and is under investigation. Both trains of Braidwood Unit 1 auxiliary feedwater started automatically following the reactor trip to maintain steam generator water levels. All systems responded as expected. Steam generator power operated relief valves lifted momentarily and reseated as designed in response to the secondary transient due to the reactor trip. The main steam dump valves are in service to the main condenser to provide heat sink cooling. The plant is being maintained at normal operating pressure and temperature. AC power is being provided by offsite power with the diesel generators in standby and all safety systems available. There is no impact to Unit 2. This report is being made per 10CFR50.72(b)(2)(iv)(B) for a RPS actuation, 4-hr. notification, and per 10CFR50.72(b)(3)(iv)(A) for an automatic actuation of the auxiliary feedwater system, 8-hr. notification. All rods inserted into the core during the trip. Concerning the relief valves momentarily lifting and reseating, there is no known primary-to-secondary leakage. The licensee has notified the NRC Resident Inspector.
ENS 5337130 April 2018 14:53:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopAt 1124 CDT, Braidwood Unit 1 experienced an automatic Reactor Trip. The cause of the Reactor Trip was a Turbine Trip with reactor power greater than P-8. The turbine trip was actuated as a result of a Turbine Motoring Generator Trip. The cause of the generator trip is unknown at this time and is under investigation. After the Reactor Trip occurred, the 1A Auxiliary Feedwater pump was manually started to provide feedwater flow to all four steam generators. The 1A Auxiliary Feedwater pump was subsequently secured and placed in standby when the Startup Feedwater pump was placed in service. Train A Main Control Room Ventilation Filtration system shifted to Makeup Mode due to a spurious actuation signal. No secondary relief valves lifted and no secondary steam was released as a result of the Reactor Trip. The Main Steam dump valves are in service to the Main Condenser to provide heat sink cooling. The plant is being maintained at normal operating pressure and temperature. AC power is being provided by Offsite Power with the Diesel Generators in standby and all safety systems available. There is no impact to Unit 2. This report is being made per 10 CFR 50.72(b)(2)(iv)(B) for a RPS actuation, 4-hr notification, and per 10 CFR 50.72(b)(3)(iv)(A) for a manual actuation of the Auxiliary Feedwater system, 8-hr notification. The licensee notified the NRC Resident Inspector and Illinois Emergency Management Agency.
ENS 5335822 April 2018 22:40:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopOn Sunday, April 22, 2018 at 1646 CDT, a valid actuation of Engineered Safety Feature (ESF) Bus 141 Undervoltage (UV) Relay occurred. At the time, Braidwood Station Unit 1 was performing a pre-planned 1A Diesel Generator (DG) Emergency Core Cooling System (ECCS) Actuation Surveillance, initiating the 1A DG to emergency start and sequence loads on a safety injection signal. Following the 1A DG solely supplying electrical power to Bus 141, the 1A DG lost voltage, resulting in an unplanned UV actuation of ESF Bus 141. The 1A DG output breaker was manually opened and local emergency stop of the 1A DG was attempted. The 1A DG continued to run at idle. Fuel supply was secured to the 1A DG and the engine stopped. Subsequently, operators restored power to ESF Bus 141 from the Unit 1 Offsite Power Source. Shutdown cooling was maintained throughout the event as the 1B Residual Heat Removal train was unaffected by the actuation. This event is reportable under 10 CFR 50.72(b)(3)(iv)(A) for 'Any event or condition that results in valid actuation of any of the systems listed...', specifically 10 CFR 50.72(b)(3)(iv)(B)(8) for the 'Emergency ac electrical power systems, including: emergency diesel generators (EDGs)...'. The licensee notified the NRC resident inspector.
ENS 5335420 April 2018 22:22:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopOn Friday, April 20, 2018 at 1730 CDT, during the Braidwood Station Unit 1 refueling outage (A1R20), a scheduled ultrasonic test (UT) was performed on the top head to upper center disc weld of the Unit 1 reactor head. The UT identified 19 indications, 9 of which are not acceptable per ASME Section XI, 2001 Edition, 2003 Addenda, Paragraph IWB-3510. This event is reportable under 10 CFR 50.72(b)(3)(ii)(A) for 'Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded'. The licensee notified the NRC Resident Inspector.
ENS 5335320 April 2018 17:57:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopOn Friday, April 20, 2018 at 1042 CDT, Braidwood Station Unit 1 was at 0 percent power in Mode 6. The 1A Diesel Generator (DG) was inoperable with troubleshooting in progress. The 1B DG was being run for a normal monthly run in accordance with 1 BwOSR 3.8.1.2-2, 'Unit One 1B Diesel Generator Operability Surveillance,' and subsequently tripped. The trip was due to a failure of the overspeed butterfly valve actuator and springs, and not an actual overspeed condition. The unit entered Technical Specification (TS) 3.8.2, 'AC Sources - Shutdown,' Condition B for required DG inoperable. All required TS actions were met at the time of the 1B DG inoperability. The offsite power source remains available. At no time was residual heat removal lost. This event is reportable under 10 CFR 50.72(b)(3)(v)(B) for any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to remove residual heat. The licensee has notified the NRC Resident Inspector.
ENS 5334719 April 2018 20:00:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopOn Thursday, April 19, 2018 at 1152 CDT, a valid actuation of Engineered Safety Feature (ESF) Bus 141 Undervoltage (UV) Relay occurred. At the time, Braidwood Station Unit 1 was performing a pre-planned Bus 141 Undervoltage Actuation Surveillance, initiating the 1A Emergency Diesel Generator (EDG) to emergency start and sequence loads on the UV signal. Following the 1A EDG solely supplying electrical power to Bus 141, the EDG lost voltage resulting in an unplanned UV actuation of the ESF Bus 141. Subsequently, operators restored power to ESF Bus 141 via crosstie of the Unit 2 offsite power source. Shutdown cooling was maintained throughout the event as the 1B Residual Heat Removal train was unaffected by the actuation. This event is reportable under 10 CFR 50.72(b)(3)(iv)(A) for 'Any event or condition that results in valid actuation of any of the systems listed...', specifically 10 CFR 50.72(b)(3)(iv)(B)(8) for the 'Emergency ac electrical power systems, including: emergency diesel generators (EDGs)...'. The licensee notified the NRC Resident Inspector.
ENS 5318023 January 2018 05:02:00BraidwoodNRC Region 3Westinghouse PWR 4-Loop

At 0400 (CST) on 1/23/2018 the Braidwood Technical Support Center (TSC) HVAC (Heating, Ventilation and Air Conditioning) Emergency Makeup Air Filter train was taken out of service to perform a planned Makeup Air Filter charcoal replacement. The TSC HVAC Makeup Air Filter train will be rendered nonfunctional during the charcoal replacement. Subsequent charcoal and HEPA filter testing will restore functionality of the TSC HVAC Makeup Air Filter train. The expected duration of the charcoal replacement and subsequent testing is 30 hours. If an emergency is declared requiring TSC activation during the time TSC HVAC is non-functional, the TSC will be staffed and activated using existing emergency planning procedure unless the TSC becomes uninhabitable due to ambient temperature, radiological, or other conditions. If relocation of the TSC becomes necessary, the Emergency Director will relocate the TSC staff to an alternate location in accordance with applicable site procedures. This notification is being made in accordance with 10 CFR 50.72(b)(3)(xiii) due to a major loss of emergency preparedness capability. An update will be provided once the TSC HVAC Emergency Makeup Air Filter train functionality has been restored. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE AT 1645 EST ON 01/26/2018 FROM PAUL ARTUSA TO JEFF HERRERA * * *

On 1/26/18 at time 1539 EST, the TSC HVAC Emergency Makeup Air Filter train was returned to service following the planned Makeup Alr Filter charcoal replacement. Functionality was verified by charcoal and HEPA filter post maintenance testing. The licensee has notified the NRC Resident Inspector. Notified the R3DO (Cameron).

ENS 5280212 June 2017 13:50:00BraidwoodNRC Region 3Westinghouse PWR 4-Loop

On Sunday, June 11, 2017 at 1200 CDT, a Chemistry Technician reported water from the Circulating Water Blowdown (CW B/D) valve building had been pumped onto the ground outside the CW B/D building during a maintenance activity. The sump in the CW B/D building has a sump pump discharge hose that had been routed outside to the ground near the outfall canal rather than to the permitted outfall path. Water sample analysis results confirmed the presence of tritium. The sump pump discharge hose was rerouted to the required outfall on June 11th, 2017 at 1500 CDT. The discharge flow did not leave the site boundary before reaching the permitted discharge pathway. In accordance with the Braidwood Station Illinois Environmental Protection Agency (IEPA) Consent Order dated March 11, 2010, the IEPA was notified of the blowdown line release on June 12, 2017 at 0914 CDT; the Illinois Emergency Management Agency (IEMA) was subsequently notified. Local agencies in the Braidwood area will also be notified. Follow-up analyses are being performed to determine if additional IEPA or IEMA notifications will be required. The release tank values for other radionuclides (i.e., other than tritium) were interpolated based on tritium values for the release tank, dilution from the CW B/D, and sampled sump tritium values; and determined to be less than Minimum Detectable Activity (MDA). Within 4 hours of determination that state and local agencies notification will be made, the NRC Operations Center is required to be notified. There were no public health risks associated with the event. The NRC Resident Inspector was notified. Due to the notification of government agencies, this event is being reported under 10 CFR 50.72(b)(2)(xi).

  • * * UPDATE ON 6/14/17 AT 1650 EDT FROM JOHN LOGAN TO DONG PARK * * *

The Illinois Environmental Protection Agency (IEPA) and the Illinois Emergency Management Agency (IEMA) were notified on Wednesday, June 14, 2017, at 1206 CDT and 1213 CDT, respectively, of the preliminary results of estimated quantity of curies of tritium release (approximately 0.009 curies of tritium) and the estimated volume of the release (approximately 35,000 gallons). The discharge flow did not leave the site boundary before reaching the permitted discharge pathway. The NRC Resident Inspector was notified of the update to IEPA and IEMA. Notified R3DO (Stone).

  • * * UPDATE FROM CRAIG FOBERT TO HOWIE CROUCH AT 1656 EDT ON 6/16/17 * * *

The Illinois Environmental Protection Agency (IEPA) and the Illinois Emergency Management Agency (IEMA) were provided with a follow-up report on June 16, 2017, at 1528 CDT as required by 35 IAC 1010.204. The report includes the estimated quantity of curies of tritium released (approximately 0.009 curies), the estimated volume of the release (approximately 35,000 gallons), and actions taken in response to the on-site release (installation of groundwater monitoring wells, implementation of a sampling program to monitor the concentrations and migration, if any, of the tritium in the groundwater on-site, and the installation of a remediation system). The NRC Resident Inspectors have been notified of the update to IEPA and IEMA. Notified R3DO (Stone).

  • * * UPDATE ON 6/20/17 AT 1435 EDT FROM MIKE NOLAN TO BETHANY CECERE * * *

On 6/20/2017, the Illinois Environmental Protection Agency (IEPA) Des Plaines Regional Office was sent a follow-up report pursuant to the Consent Order, dated March 11, 2010. This report contains that same information as the follow-up report required by 35 IAC 1010.204 that was submitted to the IEPA and the Illinois Emergency Management Agency on June 16. 2017. The NRC Resident Inspector has been notified of the update to IEPA. Notified R3DO (Pelke).

ENS 522753 October 2016 00:02:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopOn October 2, 2016 during the Braidwood Station Unit Refueling outage (A1R19), an in-service Liquid Penetration examination was performed on the previously repaired control rod drive mechanism (CRDM) penetration 69. During the examination on the weld build up for CRDM penetration 69, two indications were discovered. A 7/32 inch rounded indication was discovered located at 359 degrees on the reactor head portion of the weld buildup, and it is 4 inches from the transition of the head to penetration. A 1/4 inch rounded indication was also discovered located at 200 degrees at the transition of the head to penetration. 0 degrees is located at the outermost portion of the penetration on the flange side. The transition is the point where the vertical portion of the penetration meets the horizontal area of the reactor head. Rounded indications that exceed 3/16 inch are rejectable per ASME Code Case N-729-1. This is reportable pursuant to 10CFR50.72(b)(3)(ii)(A), 'Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded' since the as found indication did not meet the applicable acceptance criteria referenced in ASME Code Case N-729-1 to remain in-service without repair. The NRC Resident Inspector has been informed.
ENS 5210318 July 2016 23:20:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopLoss of the Offsite Emergency Response Organization (ERO) notification system (Everbridge) identified the system cannot notify all ERO individuals. This constitutes a loss of offsite communications capability. The Offsite ERO notification system (Everbridge) capability loss of Braidwood was identified at approximately 1500 CDT on July 18, 2016, due to an undetermined loss of system communications, which is currently being investigated. The Offsite ERO notification system (Everbridge) capability loss for the common Emergency Response Facility (ERF) (Emergency Offsite Facility (EOF) at Cantera) was identified at approximately 1500 CDT on July 18, 2016. The issue has subsequently been reported resolved by the vendor at 1912 CDT and site testing has verified resolution at 2108 CDT. The onsite communication system was not affected. This event is reportable under 10 CFR 50.72(b)(3)(xiii) as a loss of communications capability. The NRC Resident Inspector has been notified.
ENS 5195925 May 2016 17:10:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopOn May 25, 2016, during evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, Braidwood Station identified non-conforming conditions in the plant design such that specific TS equipment on both units is considered to not be adequately protected from tornado missiles. Each of the following reportable conditions is postulated by tornado missiles entering openings through the wall that separates the Auxiliary Building and the Turbine Building: The following items could be impacted by tornado missiles entering the 0A and 0B (common) MCR (main control room) turbine building intake openings: Main control room pressurization, main control room ductwork and dampers, chilled water to the VC (control room ventilation) coils, main control room radiation monitors. The following items could be impacted by tornado missiles entering the Division 11 and 21 MEER (miscellaneous electrical equipment rooms) rooms: Exhaust from the MEER (affects room cooling and MCR pressure), exhaust from a battery room, conduits and cabinets associated with the battery chargers and DC bus. The following items could be impacted by tornado missiles entering the Division 12 and 22 MEER rooms: Exhaust from the MEER (affects room cooling and MCR pressure), MEER supply fan and ductwork, battery room exhaust, and the instrument inverter cabinets. The Auxiliary Feed Diesel exterior exhaust stacks (which extend from the roofline) could be impacted by tornado missiles. While we have analyses on these impacts, they do not use a NRC accepted methodology to evaluate the impact and effects. The RWST (refueling water storage tank) roof access opening Bilco hatch is fabricated from sheet metal that is not designed to prevent all postulated tornado missiles from entering the tank. The tank pressure boundary is 24" thick concrete and is designed to withstand an external tornado missile impact. Thus a missile that enters the tank will not adversely impact the tank pressure boundary. The following items could be impacted by tornado missiles entering the RWST roof access: The 6" RWST recirculation pipe, 3" overflow pipe, and 24" suction pipe. This piping is located inside the tank and they are approximately 130 degrees around the tank away from the hatch opening. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) for any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety, and per 10 CFR 50.72(b)(3)(v)(D) for any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. These conditions are being addressed in accordance with EGM 15-002 and DSS-ISG-2016-01 (enforcement discretion and interim guidance documents). The licensee has notified the NRC Resident Inspector.
ENS 519129 May 2016 17:18:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopAt approximately 1420 CDT on 5/9/2016, the Braidwood Station Main Control Room was notified by Exelon Corporate Emergency Preparedness of the inadvertent actuation of all 27 Full Sounding Sirens in Will County, Illinois. Exelon Emergency Preparedness was notified at 1309 CDT on 5/9/2016. The inadvertent actuation occurred on 5/7/16 at 2350 CDT during a training activity performed by Wescom (Western Will County Communications Center). This event is reportable per 10 CFR 50.72(b)(2)(xi), News Release or Notification of Other Government Agencies. This is a 4-hour reporting requirement. The Braidwood NRC Resident has been notified.
ENS 517717 March 2016 02:00:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopThe Auxiliary Feedwater (AF) system at Braidwood automatically supplies feedwater to the Steam Generators (SG) to remove decay heat from the Reactor Coolant System following a loss of normal feedwater supply. The AF System consists of a motor driven pump (A) and a diesel driven pump (B) configured into two trains for each unit. Each pump provides 100% of the required AF capacity to the SGs as assumed in the accident analysis. One pump at full flow is sufficient to remove decay heat and cool the unit to Residual Heat Removal (RHR) entry conditions. The diesel driven AF pump is powered from an independent diesel whose combustion air intake is located in the Seismic Category II (non-seismically qualified) Turbine Building but the diesel and pump are located in the Seismic Category I (seismically qualified) Auxiliary Building. During the ongoing NRC Component Design Basis Inspection at Braidwood Station, Inspectors asked about the acceptability of the diesel combustion air intake being located in the non-seismic Turbine Building. During the review of available documentation related to the AF diesel engine combustion air intake, it was identified that the documentation did not support operation of the diesel with High Energy Line Break (HELB) environmental conditions in the Turbine Building. This has been reviewed and determined to be applicable to Braidwood Station Units 1 and 2. Specifically, prior evaluations did not account for air displacement by steam release during the event. After running different models for the Turbine Building HELB, diesel driven AF pump operability was supported for all but the Main Feedwater (FW) HELB. For the FW HELB, the best air density obtained failed to remain above the required levels deemed acceptable for engine operation and remained suppressed for extended periods of time. Additional efforts to qualify the FW piping in the Turbine Building for an Operating Basis Earthquake (OBE) to eliminate this piping from HELB considerations were not successful. This condition applies to both Units 1 and 2 but does not affect the motor driven AF pumps. This event does not constitute a loss of safety function at the point of discovery because the Braidwood opposite train motor driven AF pumps were operable on both Units 1 and 2. This event is reportable per 10 CFR 50.72(b)(3)(ii)(B) for 'any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.' The licensee has notified the NRC Resident Inspector. The licensee entered a 72-hour Action Statement and is preparing to address the issue with a configuration change.
ENS 514505 October 2015 08:41:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopBraidwood Unit 2 was performing a planned plant shutdown for refueling outage A2R18. In accordance with plant shutdown procedures while in Mode 1 (Power Operations) at approximately 15% power, operators attempted to start the Start Up Feedwater (SFWP) pump and the pump immediately tripped on Phase A Overcurrent. The 2A Motor Driven Feedwater pump (MDFWP) was manually started to maintain Steam Generator Water Level during the shutdown and subsequent plant cooldown. While in Mode 3 (Hot Standby) at (550 Degree-F), the 2A MDFWP was manually secured due to pump inboard journal bearing temperature exceeding its (200 Degree-F) operating limit. At 0105 (CDT) an anticipated automatic Auxiliary Feedwater actuation signal was generated on low Steam Generator level (36.3%) and both the 2A and 2B Auxiliary Feedwater pumps (AFP) auto-started. Also at 0105 (CDT) a Reactor Protection System (RPS) Reactor trip signal was received due to low Steam Generator level (36.3%) with the reactor not critical. Both Auxiliary Feedwater trains operated as designed with the Main Steam Dumps in service and the Main Condenser providing the heat sink. All systems operated as designed with the exception of the SFWP and the MDFWP described above. The plant is currently stable in Mode 5 with both AFPs secured. This report is being made per 10 CFR 50.72(b)(3)(iv)(A) for automatic actuation of the (1) RPS Reactor Trip with the reactor not critical and (6) Auxiliary Feedwater System, 8 hour notification. The licensee notified the NRC Resident Inspector.
ENS 5133420 August 2015 22:28:00BraidwoodNRC Region 3Westinghouse PWR 4-Loop

On 8/20/2015 at 1710 CDT, a design flaw was discovered with the pressurizer power operated relief valve (PZR PORV) block valve control circuitry. Specifically, the circuit deficiency for which a design basis fire in the Main Control Room (MCR) or cable spreading room could prevent the PZR PORV block valves from being closed from the local control switch at their associated motor control center (MCC). Engineering has reviewed this issue and determined that a potential fire induced ground in the MCR or cable spreading room could clear the associated control power fuses which would prevent the block valves from operating at the local control switch. These valves are considered to form a High/Low pressure interface which requires postulating a proper polarity DC cable to cable fault. Engineering has reviewed the circuit design and cable routing associated with PORVs 1(2)RY455A and 1(2)RY456 and determined that their associated cables are routed with other DC circuit cables in the MCR control board and cable spreading room raceways, such that this postulated fault could potentially cause spurious opening of one of the PORVs even after the control power fuses have been removed as directed by the station abnormal operating procedures for control room inaccessibility. This identified block valve circuit deficiency prevents the credited safe shutdown action of locally closing the block valves to mitigate the spurious operation of a PORV. Hourly fire watches of the affected MCR and cable spreading room fire zones have been implemented. In addition, the MCR is continuously staffed and the affected cable spreading room fire zones are equipped with detection and automatic suppression. This event is being reported under 10CFR50.72(b)(3)(ii)(B) for 'Any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.' The licensee has notified the NRC Resident Inspector.

  • * * UPDATE PROVIDED BY ROB SHERMAN TO JEFF ROTTON AT 1845 EDT ON 09/02/2015 * * *

During the extent of condition review, an additional design deficiency was identified with respect to the PZR PORV and PZR PORV Block valves. Specifically, the current mitigating strategy for removing PZR PORV control power fuses does not adequately prevent a PZR PORV from spuriously opening due to fire induced hot short. Furthermore, local actions to close the associated PZR PORV block valve at the motor control center (MCC) may not be effective because the MCC may not have electrical power during the design basis fire. Therefore, the credited safe shutdown action to remove the PZR PORV control power fuses does not prevent the PZR PORV from spuriously opening during design basis fires in some of the upper and lower cable spreading room fire zones. The affected Fire Zones are the same upper and lower spreading rooms previously identified and fire watches of the affected areas remain in place. The NRC Resident Inspector has been notified. Notified the R3DO (Skokowski)

ENS 5115415 June 2015 17:42:00BraidwoodNRC Region 3Westinghouse PWR 4-Loop

On 6/16/2015, planned preventive maintenance activities (will be) performed on the Braidwood Generating Station Technical Support Center (TSC), Ventilation System. The work will be completed within approximately 48 hours. This activity includes preventative maintenance that requires the TSC ventilation system to be out of service which will render the TSC ventilation system non-functional. If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures unless the TSC becomes uninhabitable due to ambient temperature or other conditions. If relocation of the TSC becomes necessary, the Emergency Director will relocate the TSC staff as necessary. This event is reportable per 10 CFR 50.72(b)(3)(xiii) for 'any event that results in a major loss of emergency assessment capability.' The planned maintenance will not be able to restore the TSC condensing unit or ventilation system to service within the facility activation time specified in the emergency plan (1 hour) in the event of an accident. The Emergency Response Organization team has been notified of the maintenance and the possible need to relocate during an emergency. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE AT 1102 EDT ON 6/18/15 FROM DAVID KORTGE TO JEFF HERRERA * * *

Braidwood Generating Station TSC ventilation was restored to available status at 0700 CDT on June 18, 2015. The previously reported system preventative maintenance has been completed. The licensee notified the NRC Resident Inspector. Notified the R3DO (Pelke).

ENS 5109227 May 2015 10:45:00BraidwoodNRC Region 3Westinghouse PWR 4-Loop

On 5/27/15, planned preventive maintenance activities are being performed on the Braidwood Generating Station Technical Support Center (TSC), Ventilation System. The work will be completed within approximately 10 hours. This activity includes preventative maintenance that requires the TSC ventilation system to be out of service which will render the TSC ventilation system non-functional. If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures unless the TSC becomes uninhabitable due to ambient temperature or other conditions. If relocation of the TSC becomes necessary, the Emergency Director will relocate the TSC staff as necessary. This event is reportable per 10CFR50.72(b)(3)(xiii) for 'any event that results in a major loss of emergency assessment capability.' The planned maintenance will not be able to restore the TSC condensing unit or ventilation system to service within the facility activation time specified in the emergency plan (1 hour) in the event of an accident. The Emergency Response Organization team has been notified of the maintenance and the possible need to relocate during an emergency. The licensee informed the NRC Resident Inspector.

  • * * UPDATE FROM PETER MOODIE TO VINCE KLCO ON 5/27/2015 AT 2153 EDT * * *

Planned work is complete and the Technical Support Center was restored to service on May 27, 2015 at 1808 CDT. The licensee will notify the NRC Resident Inspector. Notified the R3DO (Lipa).

ENS 5108926 May 2015 15:20:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopOn May 26, 2015, at approximately 1125 CDT, an empty beer can was found by a mechanical maintenance worker performing scaffolding work. Site Security took possession of the can and it was removed from the protected area. The can appears to be extremely old (20 plus years) with dust and debris on it. This report is submitted pursuant to 10 CFR 26.719(b)(1) based on the presence of an alcohol container within the protected area. The licensee notified the NRC Resident Inspector.
ENS 509576 April 2015 09:09:00BraidwoodNRC Region 3Westinghouse PWR 4-Loop

On 04/06/2015, planned preventive maintenance activities are being performed on the Braidwood Generating Station Technical Support Center (TSC) Ventilation System. The work will be completed within approximately 42 hours. This activity includes preventive maintenance on the TSC condensing unit which affects the TSC ventilation. During the planned maintenance, the TSC condensing unit will be rendered non-functional. If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures unless the TSC becomes uninhabitable due to ambient temperature or other conditions. If relocation of the TSC becomes necessary, the Emergency Director will relocate the TSC staff as necessary. This planned maintenance will not impact the emergency filtration capability of the TSC. This event is reportable per 10CFR50.72(b)(3)(xiii) for 'any event that results in a major loss of emergency assessment capability.' The planned maintenance will not be able to restore the TSC condensing unit to service within the facility activation time specified in the emergency plan (1 hour) in the event of an accident. The Emergency Response Organization team has been notified of the maintenance and the possible need to relocate during an emergency. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE FROM GREG HARNOIS TO JOHN SHOEMAKER AT 1013 EDT ON 4/8/15 * * *

The TSC ventilation system has been restored to normal operation as of 0600 CDT on April 8, 2015. The NRC Resident Inspector has been notified. Notified R3DO (Skokowski).

ENS 509533 April 2015 19:37:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopOn April 3, 2015, during the Braidwood Station Unit 1 refueling outage (A1R18), an in-service Liquid Penetration examination was performed on the previously repaired control rod drive mechanism (CRDM) penetration 69. During the examination on the weld build up for CRDM penetration 69, a 3/8 inch rounded indication was discovered located at 0 degrees on the reactor head portion of the weld build up, and it is 4 inches from the transition of the head to penetration. 0 degrees is located at the outermost portion of the penetration on the flange side. The transition is the point where the vertical portion of the penetration meets the horizontal area of the reactor head. Rounded indications that exceed 3/16 inch are rejectable per ASME Code Case N-729-1. This is reportable pursuant to 10CFR50.72(b)(3)(ii)(A), 'Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded' since the as found indication did not meet the applicable acceptance criteria referenced in ASME Code Case N-729-1 to remain in-service without repair. The NRC Resident Inspector has been informed.
ENS 5093327 March 2015 17:12:00BraidwoodNRC Region 3Westinghouse PWR 4-Loop

Braidwood Generating Station has completed a review of seismic monitor performance. The emergency preparedness plan requires seismic monitoring instruments to diagnose an earthquake for emergency action levels (EAL) HA4 (natural and destructive phenomena affecting vital areas) or HU4 (natural and destructive phenomena affecting the protected area). Contrary to that requirement, this review identified 5 times in the past 3 years that the seismic monitor was non-functional such that emergency classification at the ALERT or UNUSUAL EVENT level could not be obtained with site instrumentation. The seismic monitor is currently functional; however, the seismic monitor was determined to be non-functional on the following dates:

1. April 24, 2012
2. December 5, 2012
3. December 20, 2012
4. June 17, 2013
5. October 8, 2014

These non-functional conditions of the seismic monitor have been corrected and were entered into the Braidwood Corrective Action Program. While Exelon procedural direction allowed the use of offsite sources to obtain seismic data when the seismic monitor is incapable of assessing emergency plan Emergency Action Levels (EALs), this was not explicitly referenced in the Braidwood approved EALs. The loss of assessment capability is reportable to the NRC within 8 hours of discovery in accordance with 10 CFR 50.72(b)(3)(xiii). This report is required per 10 CFR 50.72(a)(1)(ii) as an event that occurred within 3 years of the date of discovery. Corrective actions are in progress. The licensee has notified the NRC Resident Inspector.

ENS 5068516 December 2014 17:07:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopThe licensee discovered evidence of prohibited material inside the protected area. The material has been removed. The licensee has notified the NRC Resident Inspector.
ENS 504316 September 2014 13:17:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopAt 0613 (CDT) on 9/6/2014, Technical Support Center roof mounted condensing units were found tripped by operations personnel. An attempt was made to restart both condensing units, neither condensing unit would restart. This caused a loss of Technical Support Center cooling capability. Technical Support Center Air Handling system is in service and required filtration remains available. Wet bulb temperature of the TSC, taken at 0950 (CDT) is 78.0 degrees F. Corrective action process has been initiated. This event is reportable under 10 CFR 50.72(b)(3)(xiii) as described in NUREG-1022, Rev. 3, since this condition affects an emergency response facility. The NRC Resident Inspector has been notified.
ENS 5039725 August 2014 04:10:00BraidwoodNRC Region 3Westinghouse PWR 4-Loop

On 8/25/2014, planned preventive maintenance activities are being performed on the Braidwood Generating Station Technical Support Center (TSC) Ventilation System. The work will be completed within approximately 42 hours. This activity includes preventive maintenance on the TSC condensing unit which affects the TSC ventilation. During the planned maintenance, the TSC condensing unit will be rendered non-functional. If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures unless the TSC becomes uninhabitable due to ambient temperature or other conditions. If relocation of the TSC becomes necessary, the Emergency Director will relocate the TSC staff as necessary. This planned maintenance will not impact the emergency filtration capability of the TSC. This event is reportable per 10CFR50.72(b)(3)(xiii) for 'any event that results in a major loss of emergency assessment capability.' The planned maintenance will not be able to restore the TSC condensing unit to service within the facility activation time specified in the emergency plan (1 hour) in the event of an accident. The Emergency Response Organization team has been notified of the maintenance and the possible need to relocate during an emergency. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE AT 1526 EDT ON 08/29/14 FROM ANNE MATHEWS TO S. SANDIN * * *

Braidwood Generating Station TSC ventilation was restored to available status at 1200 CDT on August 29th, 2014. The previously reported system preventative maintenance has been completed. The licensee informed the NRC Resident Inspector. Notified R3DO (Stone).

ENS 5030123 July 2014 21:32:00BraidwoodNRC Region 3Westinghouse PWR 4-Loop

At 1943 CDT on 7/23/14, Braidwood Station declared Unusual Event HU1 due to gunshots fired within the Owner Controlled area (OCA). This was a security condition without a Hostile Action. Both Units remain in Mode 1, 100% reactor power throughout the event. Local Law enforcement was contacted and investigated. Security stood down from the Security Condition at 1956 CDT. The Control Room was informed at 1930 CDT by Security of fourteen (14) gunshots heard inside the OCA. The licensee informed State and local agencies and the NRC Resident Inspector. Notified other FEDS (FEMA Ops Center, DHS NICC Watch Officer, DHS SWO) and (Nuclear SSA, FEMA NWC) via email.

  • * * UPDATE AT 2311 EDT FROM JOE CONQUEST TO S. SANDIN * * *

Braidwood Station Terminated an Unusual Event, HU1 at 2134 CDT. Local Law Enforcement Agency called an 'AII Clear' and station security restored the normal security posture at 1956 CDT. The licensee informed State and local agencies and the NRC Resident Inspector. Notified R3DO (Orth), NRR (Lund) and IRD (Gott) via email. Notified other FEDS (FEMA Ops Center, DHS NICC Watch Officer, DHS SWO, Nuclear SSA, FEMA NWC) via email.

ENS 5028215 July 2014 17:57:00BraidwoodNRC Region 3Westinghouse PWR 4-Loop

At 0942 (CDT) on 7/15/2014, the 2A Containment Spray chemical additive flow was found out of tolerance low during surveillance testing. This resulted in an unanalyzed condition in that insufficient chemical additive flow might have resulted in lower than assumed containment spray pH values during past periods. Based on the above, this is being reported in accordance with 10CFR50.72(b)(3)(ii)(B). Actions are in progress to restore the 2A Containment Spray chemical additive flow to within tolerance. The 2B Containment Spray system is operable per Technical Specification 3.6.6 and is capable of providing required chemical additive flow. The required flow is 18 to 67 gallons per minute (gpm), however, the measured flow was 17.96 gpm. The licensee has notified the NRC Resident Inspector.

  • * * RETRACTION FROM JAMES PETTY TO JOHN SHOEMAKER AT 1729 EDT ON 8/1/14 * * *

The purpose of this report is to retract ENS report #50282 (July 15, 2014). This ENS report was made for the 2A Containment Spray chemical additive flow which was found out of tolerance low during surveillance testing. At the time of reporting, it was concluded that this was an unanalyzed condition in that insufficient chemical additive flow may have resulted in lower than assumed containment spray pH values during past periods. This was reported in accordance with 10CFR50.72(b)(3)(11)(B) as an unanalyzed condition that significantly degrades plant safety. On Wednesday, July 29, 2014, Braidwood Generating Station concluded that the prior ENS notification could be retracted based on the completion of Engineering Change 398884, 'Evaluation Of 2A CS NaOH Spray Additive Test Results And Discussion of IRs 1682209 and 1683413.' The Engineering Change concluded that the approval of the alternate source term (AST) license amendment resulted in the elimination of a minimum containment spray (CS) spray pH value. The current containment release analysis does not credit the addition of sodium hydroxide (NaOH) to CS spray for fission product removal from the containment atmosphere. The long-term retention of captured fission products in the sump water assumes the sump water pH is greater than 7. This is established by the transfer of the containment spray additive tank (CSAT) contents to the sump during CS system operation. To transfer the maximum CSAT inventory to the sump within 8 hours, a minimum NaOH eductor flow of approximately 10 gpm is required. The minimum NaOH injection flow for the 2A CS eductor system exceeded 10 gpm so the eductor injection flows meet the criteria to transfer CSAT inventory to the containment recirculation sump within the expected minimum CS system operating time. The out of tolerance flow values recorded at the time of the initial ENS notification are acceptable. The licensee has notified the NRC Resident Inspector. Notified the R3DO (Lara).

ENS 502471 July 2014 12:14:00BraidwoodNRC Region 3Westinghouse PWR 4-Loop

At approximately 0715 (CDT) on 7/1/2014, during the daily morning test, Braidwood Generating Station experienced Offsite Alert and Notification System (ANS) failure of sirens BD10 and BD13 due to a loss of power to the sirens. Loss of power to the sirens was due to the adverse weather conditions experienced on the evening of 6/30/2014. Siren BD10 provides coverage for 20% of Braidwood's Emergency Planning Zone (EPZ) population and siren BD13 provides coverage for 14.1% of the Braidwood EPZ population. This event is being reported in accordance with 10CFR50.72(b)(3)(viii) for 'any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of control room indication, Emergency Notification System, or offsite notification system).' The failure of the sirens resulted in the loss of the capability to alert a large segment of the population in the EPZ (~34%) for more than 1 hour. Per NEI 13-01, Rev 0, a large segment of the population in the EPZ should be taken to mean approximately 25% of the total EPZ population. Immediate actions taken included placing a portable generator online to restore siren BD10 at 1015 CDT on 7/1/2014. ComEd is currently in the area attempting to restore normal power. Proposed long term corrective actions including an upgrade to the sirens to contain a battery backup are being evaluated. The licensee has notified the NRC Resident Inspector. The county emergency managers (for Will County - (for siren) BD10 and Grundy County - (for siren) BD13) are aware of the siren outage and the requirements to initiate mobile route alerting (FEMA approved backup) if required.

  • * * UPDATE FROM MURTAZA ABBAS TO DANIEL MILLS AT 1756 EDT ON 07/02/2014 * * *

Siren BD10 was restored to functional on 7/1/2014 at 1130 CDT. Siren BD13 was restored to functional on 7/2/2014 at 1145 CDT. The NRC Resident Inspector has been notified of this ENS update. Notified R3DO (McCraw)

ENS 5022725 June 2014 22:59:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopAt approximately 1555 CDT on Wednesday, June 25, 2014, during a review of several station abnormal operating procedures for actions related to low ultimate heat sink (UHS) level, it was discovered that the procedures do not incorporate design assumptions for shutting down the non-essential service water (WS) pumps following a loss of cooling lake dike. The pumps that take suction from the UHS include the WS, circulating water (CW) and fire protection (FP) pumps. Based on current procedural guidance, the only pumps that are secured due to a low ultimate heat sink level are the circulating water (CW) pumps based on low net positive suction (NPSH) Failing to secure the non-essential pumps on a loss of cooling lake dike failure significantly reduces the 30 day design basis UHS volume to approximately 4 days. This event is being reported in accordance with 10CFR50.72(b)(3)(ii)(B) for 'any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.' Immediate actions taken included issuance of an operations standing order providing direction to secure non-essential pumps on the failure of the cooling lake dike. Proposed corrective actions include procedure revisions. The licensee has notified the NRC Resident Inspector.
ENS 5018911 June 2014 22:42:00BraidwoodNRC Region 3Westinghouse PWR 4-Loop

At 1434 (CDT) on 06/11/14, the 1A Containment Spray chemical additive flow was found out of tolerance low during surveillance testing. This resulted in an unanalyzed condition in that insufficient chemical additive flow might have resulted in lower than assumed containment spray Ph values during past periods. The impact of the unanalyzed condition has not been fully evaluated. Based on the above, this is being reported in accordance with 10CFR50.72(b)(3)(ii)(B). Engineering analyses are in progress to evaluate the past condition. Actions are in progress to restore the 1A Containment Spray chemical additive flow to within tolerance. The 1B Containment Spray system is operable per Tech Spec 3.6.6 and is capable of providing required chemical additive flow. The required flow is 30 to 63 gpm. Measured flow was 27 gpm. The last measurement was six years ago. The licensee notified the NRC Resident Inspector.

  • * * RETRACTION PROVIDED BY JOHN LOGAN TO JEFF ROTTON AT 1803 EDT ON 06/13/2014 * * *

The purpose of this report is to retract EN report #50189 (June 11, 2014). This EN report was made for the 1A Containment Spray chemical additive flow which was found out of tolerance low during surveillance testing. At the time of reporting, it was concluded that this was an unanalyzed condition in that insufficient chemical additive flow may have resulted in lower than assumed containment spray pH values during past periods. This was reported in accordance with 10CFR50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. At approximately 1700 CDT on Thursday, June 12, 2014, Braidwood Generating Station concluded that the prior EN notification could be retracted based on the completion of Engineering Change 398472, 'Evaluation of As-Found Results for 1A Containment Spray NaOH Additive Flow.' The Engineering Change concluded that the approval of the alternate source term (AST) license amendment resulted in the elimination of a minimum containment spray (CS) spray pH value. The current containment release analysis does not credit the addition of sodium hydroxide (NaOH) to CS spray for fission product removal from the containment atmosphere. The long-term retention of captured fission products in the sump water assumes the sump water pH is greater than 7. This is established by the transfer of the containment spray additive tank (CSAT) contents to the sump during CS system operation. To transfer the maximum CSAT inventory to the sump within 8 hours, a minimum NaOH eductor flow of approximately 10 gpm is required. The minimum NaOH injection flow for the 1A CS eductor system (27.0 ' as-found' and 27.7 'as-left' gpm) exceeded 10 gpm so the eductor injection flows meet the criteria to transfer CSAT inventory to the containment recirculation sump within the expected minimum CS system operating time. The out of tolerance flow values recorded at the time of the initial notification are acceptable. The licensee has notified the NRC Resident Inspector. Notified R3DO (Daley).

ENS 501621 June 2014 17:51:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopAt 1314 (CDT) on 6/01/2014, Technical Support Center ventilation alarm was received in the main control room. The Equipment Operator reported that the trouble alarm for the roof mounted condensing units was in alarm and the condensing units were tripped. Upon resetting the alarms the condensing units ran for three to five minutes and tripped again. This caused a loss of Technical Support Center cooling capability. Technical Support Center Air Handling system and filtration remain in operation. The room temperature is being monitored locally. Wet bulb temperature at 1500 (CDT) was reported to be 78.5 deg F. Corrective action process has been initiated. This event is reportable under 10CFR50.72(b)(3)(xiii) as described in NUREG-1022, Rev. 3 since this condition affects an emergency response facility. The licensee has notified the NRC Resident Inspector.
ENS 4972314 January 2014 18:33:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopOn 10/16/2013 at 14:20 (CDT), the Technical Support Center (TSC) Emergency Makeup Filter Flow Control Damper (OVV145Y) was discovered degraded and non functional during planned maintenance activities. The degraded damper adversely impacted the TSC ventilation system function. On 10/16/2013 at approximately 15:00 (CDT), administrative actions were developed and briefed to isolate instrument air to the degraded nonfunctional damper OVV 145Y providing the capability to restore TSC Emergency Makeup Filter functionality. Isolating instrument air to the degraded damper places the damper in the failed open position, thereby restoring TSC Emergency Makeup Filter functionality. On 10/16/2013 at 20:25 (CDT), instrument air was isolated to the degraded damper OVV 145Y under administrative controls restoring TSC Emergency Makeup Filter functionality. This action was taken after determining no adverse affects on system operation. On 10/30/2013, degraded damper OVV145Y was repaired, instrument air restored and post maintenance testing completed, thereby restoring the degraded TSC emergency Makeup Filter Unit flow control damper functionality. This event is reportable per 10CFR50.72(b)(3)(xiii) for a major loss of emergency assessment capability because the emergent degraded TSC flow control damper OVV145Y adversely impacted the function of TSC Emergency Makeup Filter and was not restored within the TSC activation time (60 minutes). This event was originally determined not reportable because the capability to restore within the TSC activation time (60 minutes) existed from the time of discovery. Upon further review it was determined to be reportable because the degraded damper 0VV145Y was not restored to service within the TSC activation time (60 minutes). The licensee has notified the NRC Resident Inspector.
ENS 4937120 September 2013 14:40:00BraidwoodNRC Region 3Westinghouse PWR 4-Loop

On September 9, 2013, during the A1R17 Braidwood Station Unit 1 refueling outage, the as-found leakage of the controlled leakage devices (1RH01SA and 1RH01SB) for the safety injection (SI) system ECCS sump containment isolation valves (1SI8811A and 1SI8811B) were determined to not be 'leak tight' as described in the UFSAR. Since there was only minor leakage from the isolation valves or the associated residual heat (RH) system piping (1-2 drops/month from 1SI8811A and no leakage from 1SI8811B), there was no actual impact on offsite dose or long-term ECCS operation. However, further evaluation has concluded that there was a potential to exceed the assumed leakage limits of the Alternate Source Term (AST) calculation. The RH system is classified as a closed system outside containment meaning the system is designed to accommodate a single active failure (i.e., the failure of the 1SI8811B valve to close) and still maintain an adequate isolation barrier to release recirculation water outside containment. The encapsulation device is intended to capture and limit leakage from a potential leak in the 1SI8811A/B or piping. The controlled leakage device is built to the same standards as the remainder of the RH system recirculation water outside containment. The design function is to limit potential offsite dose due to leakage of recirculation water outside containment. This is not a specified safety function and there are no Technical Specification requirements for these devices. The encapsulation devices do not perform a containment function and are not a principle safety barrier. As there was only minor ECCS system leakage at the time of discovery, there was no impact on past offsite dose or long-term ECCS operation. This is reportable pursuant to 10 CFR 50.72(b)(3)(ii)(B) since the as-found leakage of the controlled leakage devices could have allowed RH leakage to exceed the calculated limits for ECCS systems outside containment. The NRC Resident Inspector has been informed.

  • * * RETRACTION FROM RANDY RAHRIG TO HOWIE CROUCH AT 1556 EDT ON 10/31/13 * * *

Retraction of ENS 49371 dated 9/09/2013: The purpose of this report is to retract ENS report #49371 (September 9, 2013). This report was made during Braidwood's refueling outage (A1R17) for the as-found leakage on the controlled leakage devices (1RH01SA and 1RH01SB) for the safety injection (SI) system ECCS sump containment isolation valves (1SI8811A and 1SI8811B) that were determined not to be 'leak tight' as described in the UFSAR. When the ENS notification was made on 9/9/2013, the station determined that there was a potential to exceed the assumed leakage limits of the alternate source term (AST) calculation. The ENS notification was made under 10CFR50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. At 1500 CDT on Thursday, October 17, 2013, the Braidwood Generating Station concluded that the prior ENS notification could be retracted based on the completion of Revision 2 of calculation BRW-13-0135-M, '1/2RH01SA/B Leak Rate Conversion and Test Pressure Determination'. The as-found pressure test results for the 1RH01SA and 1RH01SB valve containment assemblies would not have resulted in a total ECCS leakage outside containment in excess of that assumed in the AST dose calculation BRW-04-0038-M, 'Re-Analysis of Loss of Coolant Accident (LOCA) Using Alternate Source Terms (AST)'. The as-found valve containment assembly (VCA) pressure test results did not result in an unanalyzed condition that significantly degrades plant safety. The licensee has notified the NRC Resident Inspector. Notified R3DO (Daley).

ENS 4935318 September 2013 11:15:00BraidwoodNRC Region 3Westinghouse PWR 4-Loop

Boroscopic inspection of the reactor head assembly during refuel outage A1R17 has determined that white powder residue observed at reactor head penetrations 41, 49, 61, 65, and 73 is evidence of past Reactor Coolant System pressure boundary leakage at one or more of these locations. This determination was made today at 0510 (CDT) after completion of boroscopic inspections and consultation with the reactor head vendor. These reactor head penetrations are control rod drive mechanism (CRDM) penetrations. The location of the identified leakage is suspected to be the omega seal welded threaded connection on one or more of these CRDM penetrations. Further inspections must be performed to confirm the exact leakage location. This condition will be repaired before U-1 returns to mode 4. The licensee has notified the NRC Resident Inspector.

  • * * RETRACTION ON 9/27/13 AT 1529 EDT FROM MATTHEW HENSON TO DONG PARK * * *

The purpose of this report is to retract ENS report #49353 (September 18, 2013). This report was made after boroscopic inspection of the Unit 1 reactor head assembly during refueling outage A1R17 determined that white powder residue observed at five Reactor Pressure Vessel (RPV) head penetrations was evidence of past Reactor Coolant System (RCS) pressure boundary leakage at one or more of these locations. At 1400 CDT on Wednesday, September 25, 2013, the Braidwood Nuclear Power Station (BNPS) concluded that the prior ENS notification could be retracted because the Plant Operations Review Committee had concluded that the white powder residue observed at five Unit 1 RPV head penetrations was not evidence of past RCS pressure boundary leakage. A detailed examination plan was implemented, which included a disassembly of the Digital Rod Position Indication (DRPI) and Control Rod Drive Mechanism (CRDM) coils for all five penetrations and a detailed visual examination using remote video equipments. The visual examinations used remote video equipments to inspect the lower and intermediate canopy seal weld areas for the five CRDMs. In addition, an examination of the upper canopy seal weld areas and a bare-metal inspection of the Reactor Pressure Vessel Head did not identify any evidence of boric acid leakage. Based on these inspections and a review of available photos from previous outages, BNPS concluded that the white residues were from incomplete cleaning in prior outages. The licensee has notified the NRC Resident Inspector. Notified R3DO (Dickson).

ENS 4934314 September 2013 05:47:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopOn September 14, 2013, during the A1R17 Braidwood Station Unit 1 refueling outage, an inservice Liquid Penetration examination was performed on the previously repaired Penetration 69. This repair was performed during A1R16. The PT (Penetrant Testing) examination revealed 22 recordable indication and 5 non-relevant indications. Rounded indications that exceed 3/16" are rejectable. All 22 indications are rounded and 13 of these indications exceed the 3/16" criteria. This is reportable pursuant to 10 CFR 50.72(b)(3)(ii)(A) since the as found indication did not meet the applicable acceptance criteria referenced in ASME Code Case N-729-1 to remain in-service without repair. NRC Resident Inspector has been informed.
ENS 488853 April 2013 16:40:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopDuring remodeling of a bathroom on the third floor of the Administrative Building which is located inside the Protective Area, workers discovered two very old containers of blackberry brandy after removing the ceiling tiles. This item will be entered into the licensee corrective actions program for follow up. The licensee informed the NRC Resident Inspector. HOO Note: A similar report (EN #48877) was received on 4/2/2013.
ENS 488772 April 2013 15:52:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopDuring remodeling of a bathroom on the third floor of the Administrative Building which is located inside the Protective Area, workers discovered a very old container of gin after removing the ceiling tiles. This item will be entered into the licensee corrective actions program for follow up. The licensee informed the NRC Resident Inspector.
ENS 4857911 December 2012 12:10:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopIn accordance with 10CFR50.73(a)(2)(iv)(A), this telephone notification reports an invalid actuation of the Unit 2 train B (2B) Emergency Diesel Generator (DG) undervoltage auxiliary relay on October 19, 2012 at 2324 hours. At the time of he event, Unit 2 was in Mode 6, Refueling. The 2B DG was undergoing the completion of a monthly surveillance and was operating in parallel with the offsite power, when the 2B DG changed from the test mode to the emergency mode. The offsite feed to the bus (ACB 2422) tripped, and the 2B DG load increased to carry the bus. Proper operation of the 2B DG was verified, and expected train equipment loaded and functioned successfully. The potential cause is attributed to a spurious actuation of the undervoltage auxiliary relay. This condition was entered into the corrective action program. The licensee notified the NRC Resident Inspector and the Illinois Emergency Management Agency Resident Inspector.
ENS 4835127 September 2012 13:55:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopA licensee contractor discovered contraband material concealed above ceiling tiles in a locker room during a renovation of the area. Contact the Headquarters Operations Officer for details. The licensee will notify the NRC Resident Inspector.
ENS 4832821 September 2012 01:20:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopAt 2035 CDT on 9/20/2012 power was removed from the Technical Support Center ventilation for planned maintenance on the supply breaker and the supply breaker cubicle. At 2109 CDTduring restoration, it was discovered that the breaker for the Technical Support Center ventilation could not be closed. The cause for not being able to close the supply breaker is unknown. Troubleshooting is currently in progress and the Technical Support Center ventilation is expected to be returned to service on 09/21/2012. This event is reportable under 10 CFR 50.72(b)(3)(xiii) as described in NUREG-1022, Rev.2 since this work activity affects an emergency response facility. The licensee has notified the NRC Resident Inspector.
ENS 4793418 May 2012 12:50:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopOn April 23, 2012, during the Braidwood Station Unit 1 refueling outage, it was determined that the results of planned ultrasonic (UT) examinations performed on one penetration (Penetration No. 69) did not meet the applicable acceptance criteria, requiring repair prior to returning the reactor pressure vessel head to service. A portion of the indication was conservatively assumed to be within the J-groove weld. At the time of the examination, the Braidwood Station Unit 1 reactor pressure vessel head was classified as a low susceptibility head. The cause of the recordable indication is attributed to Primary Water Stress Corrosion Cracking. The examinations were being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D) and ASME Code Case N-729-1, to ensure the structural integrity of the reactor pressure vessel head pressure boundary. All of the remaining penetrations have been examined during the current refueling outage. Repairs to Penetration No. 69 were completed prior to commencing startup. No other repairs were required. This is reportable pursuant to 10 CFR 50.72(b)(3)(ii)(A) since the as found indication did not meet the applicable acceptance criteria referenced in ASME Code Case N-729-1 to remain in-service without repair. The NRC Resident Inspector has been notified. As a result of a follow-up review, at 0600 CDT on 05/18/12, the licensee determined this was an 8 hour reportable event while Unit 1 was in Mode 4. The licensee did not determine this was a reportable event immediately after the UT was conducted on April 23, 2012. The unit is currently in Mode 3. Reactor Vessel Head Penetration No. 69 is a control rod penetration. Reactor Coolant Leakage test are satisfactory and there are no indications of boron buildup on the Reactor Vessel Head. Control rod testing has not yet been performed.
ENS 4784517 April 2012 23:12:00BraidwoodNRC Region 3Westinghouse PWR 4-Loop

At 2130 (CDT) on April 17, 2012, the Unit 1 Plant Process Computer (PPC) was removed from service for a planned replacement in the current Unit 1 Refueling Outage. The Unit 1 PPC feeds the Safety Parameter Display System (SPDS) used in the Main Control Room (MCR) and the Technical Support Center (TSC). The Unit 1 PPC also feeds the Emergency Response Data System (ERDS). The Unit 1 and Unit 2 PPCs also feed the Plant Parameter Display System (PPDS) used in the MCR, TSC and Emergency Operations Facility (EOF). Meteorological data will remain available in the MCR but not through ERDS for either Unit 1 or Unit 2. The dose assessment program will remain functional as the Unit 2 Plant Process computer will be capable of providing the necessary data through PPDS to run the program. The dose assessment program is not affected by the Unit 1 PPC being out of service. As compensatory measures, a proceduralized backup method to fax or communicate via a phone circuit applicable data to the NRC, TSC, and EOF exists. There is no impact to the Emergency Notification System (ENS) or Health Physics Network (HPN) communication systems. The new Unit 1 PPC is scheduled to be functional on April 21, 2012. However, based on the mode Unit 1 will be in, this will limit the number of points that would provide usable data. The Unit 1 PPC will be tested as mode changes occur. The Unit 1 PPC is planned to be declared functional by Mode 2. A follow-up ENS call will be made once the Unit 1 PPC is declared functional. The loss of SPDS and ERDS is a 'major loss of assessment capability' and is reportable under 10CFR50.72(b)(3)(xiii). The NRC Senior Resident Inspector and the State of Illinois (through the Illinois Emergency Management Agency Resident Inspector) have been notified of this ENS call.

  • * * UPDATE FROM JOE KLEVORN TO JOHN KNOKE AT 1035 EDT ON 05/15/12 * * *

As of 1035 EDT on May 15, 2012, the Unit 1 PPC is considered operational with respect to the Safety Parameter Display System (SPDS), Plant Parameter Display System (PPDS) and Emergency Response Data System (ERDS). Therefore, a Major Loss of Assessment Capability no longer exists on Unit 1. The EP Manager will contact the NRC Computer Center for Unit 1 to ensure the ERDS data is being satisfactorily sent to the NRC. The NRC Resident Inspector and the State of Illinois (through the Illinois Emergency Management Agency Resident Inspector) have been notified of this ENS update. The R3DO (Patty Pelke) has been notified.

ENS 476373 February 2012 22:10:00BraidwoodNRC Region 3Westinghouse PWR 4-Loop

VOLUNTARY REPORT - DESIGN VULNERABILITY IN 4.16kV BUS UNDER-VOLTAGE SCHEME

On January 30, 2012, a design vulnerability was discovered at Byron and Braidwood stations in the Engineered Safety Feature 4.16kV bus under-voltage protection scheme for Braidwood Station Units 1 and 2. Specifically a voltage unbalance created by an open circuit of either the A or C phase from the offsite grid to the System Auxiliary Transformers (SAT) is not designed to actuate the protective relays on the 4.16kV safety bus that provides isolation from the offsite grid and the automatic start and loading of the emergency onsite diesel generators.

Two under-voltage relays are provided on each 4.16kV safety bus, which are combined in a two out of two logic to generate a loss of power signal. The relays are sensing voltage between two phases (i.e., A&B and B&C). An open circuit condition on the C phase or the A phase would not satisfy the two out of two logic. This condition results in both 4.16kV safety buses remaining energized with a bus undervoltage situation and results in equipment protective devices actuating from over-current conditions.

This configuration is a non-conforming condition in that the design of the under-voltage relays and logic was intended to identify degraded grid conditions, not loss of a single phase. With an open circuit on the A or C phase from the grid to the SATs, during normal operations, operators have to diagnose the condition and manually isolate safety buses from offsite power which would automatically start and load the emergency diesel generators. During a design basis event concurrent with an open circuit on A or C phase from the grid to the SATs, analysis performed to date indicates that starting of the ECCS loads would have caused the bus voltage to decrease sufficiently to actuate the under-voltage protective relays and restore cooling with emergency onsite power without challenging fuel design limits.

The 4.16kV safety bus under-voltage protection scheme at Byron and Braidwood is believed to be a typical industry design. This design issue is being evaluated at the other Exelon stations. The results of this evaluation will be shared with the NRC. Therefore, this condition is being reported as a voluntary notification due to its potential generic industry applicability."

The licensee notified the NRC Resident Inspector.

ENS 4762225 January 2012 21:39:00BraidwoodNRC Region 3Westinghouse PWR 4-Loop

Planned preventive maintenance activities are being performed on the Braidwood Nuclear Station Technical Support Center (TSC) Ventilation System. These work activities are planned to be performed and completed expeditiously within 8 hours. This maintenance activity includes the performance of preventive maintenance on the TSC outside air supply fan unit which affects the TSC emergency filter train and air handling unit. During a portion of the time these activities are being performed, this equipment will not be available for operation. As such, the TSC Ventilation will be rendered non-functional during the performance of portions of the work activity. If an emergency condition occurs that requires activation of the Technical Support Center, during the time this work activity is being performed, it will take no more than 4 hours to return the equipment back to functional status, dependent on the stage of the work activity at the time an emergency occurs. Plans are to utilize the TSC for any declared emergency during the time this work activity is being performed as long as radiological conditions allow. This event is reportable per 10CFR50.72(b)(3)(xiii) as described in NUREG-1022, Rev. 2 since this work activity affects an emergency response facility for the duration of the maintenance. The licensee notified the NRC Resident Inspector.

  • * * UPDATE AT 1107 EST ON 01/26/12 FROM RICHARD ROWE TO S. SANDIN * * *

Braidwood Nuclear Station TSC ventilation was restored to available status at 0635 CST on January 26, 2012. The previously reported system preventative maintenance has been completed. The licensee notified the NRC Resident Inspector. Notified R3DO (L. Kozak).

ENS 473194 October 2011 19:33:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopAt 1430 CDT, 10/4/11, the licensee Main Control room received an emergency phone call for medical assistance. The caller reported that he had a person down in the auxiliary building. On-site medical response team reported the person was not breathing and not responsive. The symptoms were non-occupational. The person was transported off site via ambulance to the local hospital. At 1635, 10/4/11, the licensee was informed that the individual was declared deceased. OSHA is being notified pursuant to the requirements of 29CFR1904.39. This ENS report is being made in accordance with 50.72 (b)(2)(xi) There was no radioactive contamination involved in this event. The licensee does not plan any media or press release and have not notified any other government agencies besides OSHA. The NRC Resident Inspector will be notified.
ENS 469429 June 2011 19:03:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopThis siren affects both Dresden and Braidwood Stations. At approximately 1422 (CDT) on June 9, 2011, a Grundy County representative notified Exelon of an inadvertent siren activation. The Exelon representative contacted the siren contractor at 1423 (CDT) inquiring about siren activity in the shared Braidwood/Dresden Emergency Planning Zones. Upon review and polling of the system at 1444 (CDT), siren BD11 was identified in the area of concern with a communication failure. The siren contractor was immediately dispatched and arrived at the siren location at approximately 1520 (CDT). Upon arrival, the siren was not sounding and significant damage was identified, which was indicative of a direct lightning strike. The siren was declared out of service and repairs will begin on June 10th. The licensee notified the NRC Resident Inspector and Grundy County local authorities. See Dresden Event #46943 for same event notification.